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최성남(Sung Nam Choi),이국희(Kuk Hee Lee),문호림(Ho Rim Moon),김준우(Jun Woo Kim),김훈태(Hune Tae Kim) 대한기계학회 2021 大韓機械學會論文集A Vol.45 No.7
원자력발전소 가동중검사 기술기준의 허용기준을 초과하는 결함이 발견되면, 결함을 정비 또는 기기를 교체해야 한다. 하지만, 결함이 발견된 기기 및 배관은 파괴역학분석으로 구조적 건전성이 확인되면 정비 또는 교체 없이 계속 사용이 가능하다. OPR1000 원전의 기기 및 배관 용접부에서 허용기준을 초과한 결함을 평가하기 위해 K-IPSIE(KHNP-integrated program for structural integrity evaluation)를 개발하였다. K-IPSIE는 ASME BPVC Section XI, App. A/C/H 결함평가 절차를 프로그램으로 구현하였으며, OPR1000 원전 기기 및 배관의 비파괴 체적검사 대상 용접부의 응력을 데이터베이스로 구축하여, 원전 가동중검사 시 결함이 발견되면 신속하고 신뢰성 있는 건전성을 평가하여 원전 이용률 향상에 기여할 것으로 판단된다. The component of a nuclear power plant in which a flaw is detected during an in-service inspection should be repaired or replaced. However, if the flawed component is proved to have structural integrity during an analytical evaluation, such as fracture mechanics analysis, the component can be operated for that evaluation period of time without any repair or replacement. In this study, the KHNP-integrated program for structural integrity evaluation (K-IPSIE) is developed for the analytical evaluation of the flawed components and pipings in OPR1000 NPPs. This program integrates the stress database of the welds of OPR1000 NPPs and the flaw analysis algorithm based on ASME BPVC Section XI, Appendices A/C/H. Because K-IPSIE exactly and promptly evaluates the integrity of the flawed components and pipings in the weldings during an in-service inspection of OPR1000 NPPs, it will enhance the operating rate of NPPs, without prolonging the outage period.
원자력 주요기기 해석을 위한 자동요소망 생성프로그램 개발
장동민,김영진,최성남,서명원,장기상,Jang, Dong-Min,Kim, Yeong-Jin,Choe, Seong-Nam,Seo, Myeong-Won,Jang, Gi-Sang 대한기계학회 2000 大韓機械學會論文集A Vol.24 No.2
Fracture mechanics analysis (FMA) is an essential work for integrity evaluation of nuclear power plant. The flaws inspected by In-Service Inspection(ISI) should be confirmed by FMA for the decision of the operation status of stop or continuance. The basic data for FMA are the stress of the interested area. The purpose of this research is to develop a system which can obtain stress data efficiently based on various database. Mesh generation program generates mesh using MSC/PATRAN and provides input file for finite element analysis according to the databases (shape, dimension, transient and material). The stress data from the finite element analysis are stored to be stress database so that it can be applied to FMA. As an example, the system developed by this study is applied to pressurizer nozzle and confirmed to be a useful tool for efficient FMA.
원자력배관 건전성평가 전문가시스템 개발(1) - 평가법 제시 및 재료물성치 추론 -
김영진,석창성,최영환,Kim, Yeong-Jin,Seok, Chang-Seong,Choe, Yeong-Hwan 대한기계학회 1996 大韓機械學會論文集A Vol.20 No.2
The objective of this paper is to develop an expert system for nuclear piping integrity. This paper describes the selection methodology of integrity evalution method and the inference of material properties. To select the integrity evaluation method, the weight factor for respective material properties was obtained by the sensitivity analysis of the effect of material properties on integrity evaluation method. Subsequently the possession ratio for respective integrity evaluation method was computed, and the most appropriate integrity evaluation method for given input information is selected. In the material properties inference, stress-strain curves and J-R curves were predicted from tensile properties such as yield strength and tensile strength.
장병욱(Byungwook Jang),강석철(Sukchul Kang),김경희(Kyungheui Kim),박정선(Jungsun Park) 대한기계학회 2009 대한기계학회 춘추학술대회 Vol.2009 No.5
The fatigue fracture which occurs under repeated loads is a major concern in mechanical component design. Recently, fatigue life evaluation is getting important to guarantee reliability and safety of a product. In the fatigue analysis and design, uncertainties are caused by the variances of geometry data and applied loads, and the scatter of material properties. It makes the deterministic methods less useful. In this reason, reliability analysis concept should be incorporated to ensure fatigue safety in more sophisticated way. In this paper, fatigue crack growth life of turbine wheel which exposed to severe environments subject to high temperature and centrifugal forces is evaluated by fracture mechanics. Also the reliability analysis is accessed by the fist order second moment method and Monte Carlo simulation.
영향계수를 이용한 원자로 압력용기의 운전제한곡선 작성 : 냉각곡선
장창희,Jang, Chang-Hui 대한기계학회 2002 大韓機械學會論文集A Vol.26 No.3
During heatup and cooldown of pressurized water reactor, thermal stress was generated in the reactor pressure vessel (RPV) because of the temperature gradient. To prevent potential failure of RPV, pressure was required to be maintained below the P-T limit curves. In this paper, several methods for constructing the P-T limit curves including the ASME Sec. XI, App. G method were explained and the results were compared. Then, the effects of the various parameters such as flaw size, flaw orientation, cooldown rate, existence of chad, and reference fracture toughness, were evaluated. It was found that the current ASME Sec. XI App. G method resulted in the most conservative P-T limit curve. As the more accurate fracture mechanics analysis results were used, some of the conservatism can be removed. Among the parameters analysed, reference flaw orientation and reference fracture toughness curve had the greatest effect on the resulting P-T limit curves.